Nuclear Energy University Programs (NEUP) Fiscal Year (FY) 2016 Annual Planning Webinar

Advanced Reactor Components

(Subtopics RC-2.1 & 2.3)

William Corwin Office of Advanced Reactor Technologies

U.S. Department of Energy

August 11, 2015

Structural Materials Are Critical for Advanced Nuclear Reactors

n  Development and qualification of advanced structural materials are critical to the design and deployment of the advanced nuclear reactor systems that DOE is developing •  High and Very High Temperature Gas Cooled Reactors

(HTGRs and VHTRs) •  Sodium Cooled Fast Reactors (SFRs) •  Fluoride Salt Cooled High Temperature Reactors (FHRs)

n  Structural materials must perform over design lifetimes for pressure boundaries, reactor internals, heat transfer components, etc.

n  Performance of metallic alloys for the long times and high operating temperatures required is being examined under the Advanced Reactor Technologies (ART) Program

ART Program Supports Advanced Reactor Development

n ART R&D supports multiple high-level objectives identified in the 2010 Nuclear Energy R&D Roadmap (2 & 3) (2) Develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration’s energy security and climate change goals (3) Develop sustainable nuclear fuel cycles …overall goal is to have demonstrated the technologies necessary to allow commercial deployment of solution(s) for the sustainable management of used nuclear fuel that is safe, economic, and secure and widely acceptable to American society by 2050.”

ART Program Includes Advanced Materials R&D Activities

n  Development and qualification of graphite, improved high-temperature alloys, and ceramic composites for advanced reactor systems

n  Advanced Fast Reactor-100 is an example of fast reactor systems

•  Targets local small grids with limited needs for on-site refueling

•  250MWt/100MWe, sodium-cooled, core life (30 years), plant life (60 years)

n  AREVA’s High Temperature Reactor is an example of a He-cooled system

•  TRISO fueled, graphite moderated •  625MWt/315MWe, 750°C outlet

temperature to steam generator, plant life (60 years)

AFR-100

HTR

Advanced Reactor Components Addresses Two Significantly Different Structural Materials Research Topics in FY16

n  RC-2.1 Compact Heat Exchangers (CHXs) •  Advanced nuclear reactor systems may utilize CHXs as part of high

efficiency, Brayton cycle energy conversion systems for improved cost and efficiency. SFRs and FHRs may utilize a supercritical carbon dioxide (sCO2) secondary loop, and HTGRs may incorporate CHXs. Numerous ASME Codification issues must be addressed before CHXs can be used in nuclear systems

n  RC-2.3 Helium Tribology in Gas Cooled Reactors •  HTGRs and VHTRs use helium as a reactor coolant to transfer heat

from the core to either a primary-to-secondary heat exchanger or to a steam generator. Metallic components will likely consist of Alloy 800H or Inconel 617. Friction, wear, and self-welding of metallic surfaces in the typical He environment of such reactors is of concern.

RC-2.1 CHXs for Advanced Reactor Systems Require ASME Nuclear Code Qualification

n High efficiency Brayton cycle systems based on sCO2 improve cost and efficiency •  Minimize material cost and

reactor and plant footprint

n CHX for SFR Brayton cycle operates at moderate high temperature – about 550oC •  Nominal sCO2 pressure about

20MPa (2900psi) •  Nominal sodium pressure about

2MPa (300psi) n Potential HTGR and VHTR

applications at higher temperature (800 – 950oC)

Example of CHX modules inside a Section III vessel

Dostal et al., MIT-ANP-TR-100, 2004

Micro Channel and Plate Fin Are Candidate CHX Types

n  Diffusion bonded micro channel •  Series of bonded thin plates or shims •  Chemically etched micro channels •  Integral flow distribution channels •  Good for high pressure

n  Diffusion bonded or brazed fin CHX •  Corrugated plate fin sandwiched between two flat plates or

shims •  More efficient use of materials and larger flow passages

n  Composite of both types proposed for sCO2 CHX •  High pressure sCO2 in micro channels •  Low pressure sodium in plate fin

configuration

Southall et al., ICAPP ‘08

Sabharwall et al., Journal of Thermal Service and Engineering Applications, 2013

Southall et al., ICAPP ‘10

CHXs Have Complex Geometry at Micro- and Macro-Scales

n  Complex arrangement of flow channels and distribution manifolds •  Challenge to ensure external

boundary conditions transfer to CHX internals

n  Sharp re-entrant corners result in stress concentrations •  Significant stress redistribution

from pressure and thermal stresses

Southall et al., ICAPP ‘10

Southall et al., ICAPP ‘09

Complex CHX Structure Challenges Current ASME Nuclear Design Process

n  Non-nuclear Sec VIII, Div 1 provides micro channel sizing rules based on stayed rectangular vessels •  Does not consider thermal stresses or cyclic loading •  Does not adequately address elevated temperature

bending stress/strain redistribution due to creep n  Sec III elevated temperature design analysis

(Subsec HBB) requires consideration of thermal stresses and cyclic loading based on stress classification •  Thermal stresses important for allowable cyclic life –

particularly at elevated temperature –  Interaction of core and manifold –  Core thermal gradients – global and local

•  Stress classification requires deconstruction of FEA results (primary, secondary and peak)

•  Stress classification at sharp corner incompatible with current Section III, Subsection HBB methods

Complex thermal mechanical interactions at header location

Component SMTDesign Curve

δ test

ε ,  E test

testq

Elastically  Calculated  

Strain  Ran

ge

N Allowable  Cycles

nq

compδ

Lq

, E compε

Elastically calculated strain range vs. cycles to failure

R

Wang et al., Journal of Pressure Vessel Technology, 2014

Potential Solutions Are Available to Develop Bases for ASME Sec III Rules

n  Substucturing •  Define equivalent smeared properties surrounding detailed representation of local area •  Still requires stress classification

n  Sec III Div 5 elastic-perfectly plastic (E-PP) draft Code Cases •  Uses elastic-perfectly plastic stress analysis based on pseudo yield stress that bounds

response to allowable stress/strain values •  Doesn’t use stress classification •  Models all details, thermal gradients and cyclic loading

n  Impact of sharp corner stress concentration •  Experimentally determine effective

stress/strain concentration factor or effective corner radius

•  Experimentally establish equivalent creep-fatigue curve analogous to SMT approach

•  Creep and creep-fatigue crack growth methodologies

RC-2.1 Compact Heat Exchangers

n Proposals are sought develop a structural design methodology to provides an assessment of elevated temperature failure modes of CHX under sustained and cyclic loading due to pressure and thermal gradients. The results will be used as the technical basis for developing a Section III, Division 5 Code Case for CHXs in nuclear systems. Input from industry would be considered valuable.

n Potential solution pathways could focus on the following areas to resolve the complex CHX structure challenges needed for current nuclear design procedures •  Substructuring •  Elastic-perfectly plastic design methodology •  Experimental approach to deal with sharp stress concentrations

at corners Draft Report on CHX Nuclear Code Needs Being Finalized—Late August? J. Nestell and T.-L. Sham, “ASME Code Considerations for the Compact Heat Exchanger,” ORNL/TM-2015/401, Oak Ridge National Laboratory, Oak Ridge, TN (to be available from osti.gov)

RC-2.3 Tribology Performance of Metallic Alloys in High Temperature Helium Environments Needs to Be Examined

n  High temperature strength and environmental compatibility in HTGR helium coolant is critical

n  Creep and fatigue of ASME approved alloys appear to be relatively unaffected by HTGR helium

n  Impact of HTGR helium on tribology of metallic components is less clear –  Friction –  Wear/Fretting –  Self Welding

Nominal composition & potential ranges for HTGR He (µatm) H2 H2O CO CO2 CH4 N2 Nominal 400 2.0 40 0.20 20 6 Range 300-500 1.0-4.0 30-50 0.05-0.40 10-30 2-10

Metallic Components in HTGR Helium that Undergo Relative Motion Are Susceptible to Adverse Tribology Effects

Simplified Helium Circulator

German HTR He Isolation Valve

Prototype

HTGR Control Rod

ASME Code-Approved Alloys for High Temperature Nuclear Construction Should Be Evaluated for He Tribology

n High temperature alloys 800H and Alloy 617 (soon to be added to ASME III Div 5) are of greatest interest

n Multiple mechanisms need to be evaluated •  Friction and wear of sliding parts • Quasi-static situations (thermal expansion) •  Vibrations and translations •  Self-welding

n Example components •  Control rod drive mechanisms •  Fuel handling mechanisms for prismatic designs •  Pneumatic systems for pebble bed fuel •  Valves stems and seats (may be hardfaced) •  Helium circulators

RC-2.3 Effects of HTGR Helium Coolant on Metallic Component Tribology

n Proposals are sought to develop an improved understanding of the potential for enhanced tribological damage to Alloy 800H and 617 in simulated HTGR helium at relevant reactor operating temperatures. Input from industry involved in the design or manufacture of primary circuit components subject to helium tribology issues would be considered valuable.

n The proposed research should focus on •  Experimental studies to compare the friction, wear, and self welding

of the similar and dissimilar metal alloy couples in HTGR helium compared to that in air as functions of time, load, rate, environment, etc.

•  Conditions under which significant enhancement of deteriorated behavior in helium is observed should be identified and quantified.

•  Mitigation approaches should be developed and described, if possible.

Nuclear Energy University Programs (NEUP) Fiscal Year (FY) 2016 Annual Planning Webinar

Electromechanical Pumps

(Subtopic RC-2.2)

Carl Sink Office of Advanced Reactor Technologies

U.S. Department of Energy

August 11, 2015

Evaluating End Effects in EM Pumps

In  an  Annular  Linear  Induc.on  EM  Pump  (ALIP),  the  traveling  magne.c  field  of  the  stator  induces  a  current  in  the  fluid;  the  resultant  axial  force  on  the  fluid  induces  a  direc.onal  flow.    ASTRID    uses  “matrix  method”  to  solve  Maxwell’s  equa.ons    in  a  series  of  mesh  calcula.ons.    But  correc.on  factors  for  various  fringing  fluxes  and  stator  ends  result  in  an  empirical  outcome  that  requires  improvement.   Large  electro-­‐magne.c  pump  design  for  applica.on  in  the  ASTRID  sodium-­‐  cooled  fast  reactor.  Guy  

LAFFONT,  Frédéric  REY,  Rie  AIZAWA,  Tetsu  SUZIKI  

End effects in EM Pumps

Three  dimensional  currents  are  generated  in  regions  where  conduc.ng  fluids  flow  through  spa.ally  varying  magne.c  fields    These  3-­‐D  currents  interact  with  the  magne.c  field,  thus  producing  addi.onal  MHD  pressure  drop    Simple  methods  to  account  for  the  "end  effects”  or  pressure  losses  at  the  ends  of  the  pump,  such  as  reducing  the  calculated  value  by  a  factor  of  (N-­‐1)/(N+1),  where  N=#  of  poles,  are  no  longer  adequate  

Unpublished  results,  Sergei  Molokov,  Coventry  University.  2001  

EM Pump End Effects

Ignoring  end  effects  in  ALIPs  results  in  over-­‐es.ma.ng  pressure  rise  

Ignoring  end  effects  in  ALIPs  results  in  over-­‐es.ma.ng  pumping  efficiency  

The  magne.c  field  and  performance  calcula.ons  for  an  electromagne.c  pump  of  a  liquid  metal,  Suwon  Choy  and  Sang  Hee  Hongz,  J.  Phys.  D:  Appl.  Phys.  31  (1998)  2754–2759  

RC-2.2 Electromechanical Pumps

n The objective of this project is to analyze the impact of end effects on the magnetic field and pump efficiency of current SFR EM pump designs. The project will also consider coupling results obtained from EM pump performance analyses with modern CFD modeling techniques.